1. Field of the Invention
The present invention relates to zirconium based alloys suitable for use in nuclear reactor service, and in particular for use in claddings of fuel elements.
2. Description of Related Art
Nuclear fuel element cladding serves several purposes and two primary purposes are: first, to prevent contact and chemical reactions between the nuclear fuel and the coolant or the moderator if a moderator is present; and second, to prevent the radioactive fission products, some of which are gases, from being released from the fuel into the coolant or the moderator. The failure of the cladding, i.e., a loss of the leak-proof seal, can contaminate the coolant or moderator and the associated systems with radioactive long-lived products to a degree which interferes with plant operation.
Zirconium-based alloys have long been used in the cladding of fuel elements in nuclear reactors. A desirable combination is found in zirconium by virtue of its low thermal neutron cross-section and its generally acceptable level of resistance to corrosion in a boiling water reactor environment. Zircaloy 2, a zirconium alloy consisting of about 1.2 to 1.7 percent tin, 0.07 to 0.2 percent iron, 0.05 to 0.15 percent chromium, 0.03 to 0.08 percent nickel, up to 0.15 percent oxygen, and the balance zirconium, has enjoyed performance in reactor service, but also possesses some deficiencies that have prompted further research to find materials providing improved performance. For example, Zircaloy 2 cladding on fuel elements in nuclear reactors absorbs hydrogen while the reactor is operating. When the reactor is shut down and the cladding cools the Zircaloy 2 is embrittled by the absorbed hydrogen. Zircaloy 4 was one alloy developed as a result of further research to improve Zircaloy 2. Zircaloy 4 is similar to Zircaloy 2 but contains less nickel (0.007% max. wt. percent) and slightly more iron. Zircaloy 4 was developed as an improvement over Zircaloy 2 to reduce absorption of hydrogen in Zircaloy 2. Zircaloy 2 and Zircaloy 4 are herein referred to as the Zircaloy alloys or Zircaloy.
The Zircaloy alloys are among the best corrosion resistant materials when tested in water at reactor operating temperatures, typically about 290.degree. C., but in the absence of radiation from the nuclear fission reaction. The corrosion rate in water at 290.degree. C. is very low and the corrosion product is a uniform, tightly adherent, black ZrO.sub.2 film. In actual service, however, the Zircaloy is irradiated and is also exposed to radiolysis products present in reactor water. The corrosion resistance properties of Zircaloy deteriorate under these conditions and the corrosion rate thereof is accelerated.
Research efforts directed at improving the corrosion properties of the zirconium-based alloys have yielded some advances. Corrosion resistance has been enhanced in some instances through carefully controlled heat treatments of the alloys either prior to or subsequent to material fabrication. Added heat treatment cycles, however, generally increase the expense of making finished products, and in those instances where an installation requires welding to be performed, the area affected by the heat of the welding operation may not possess the same corrosion resistance characteristics as the remainder of the article. Variations in the alloying elements employed and the percentages of the alloying elements have also been propounded in an effort to address the deterioration in the corrosion-resistance of these alloys when they are irradiated.
The deterioration under actual reactor conditions of the corrosion resistance properties of Zircaloy is not manifested in merely an increased uniform rate of corrosion. Rather, in addition to the black ZrO.sub.2 layer formed, a localized, or nodular corrosion phenomenon has been observed in some instances on Zircaloy tubing in boiling water reactors. In addition to producing an accelerated rate of corrosion, the corrosion product of the nodular corrosion reaction is a highly undesirable white ZrO.sub.2 bloom which is less adherent and lower in density than the black ZrO.sub.2 layer.
The increased rate of corrosion caused by the nodular corrosion reaction will be likely to shorten the service life of the tube cladding, and also this nodular corrosion will have a detrimental effect on the efficient operation of the reactor. The white ZrO.sub.2, being less adherent, may be prone to spalling or flaking away from the tube into the reactor water. On the other hand, if the nodular corrosion product does not spall away, a decrease in heat transfer efficiency through the tube into the water is created when the nodular corrosion proliferates and the less dense white ZrO.sub.2 covers all or a large portion of a tube.
Actual reactor conditions cannot be readily duplicated for normal laboratory research due to the impracticality of employing a radiation source to simulate the irradiation experienced in a reactor. Additionally, gaining data from actual use in reactor service is an extremely time consuming process. For this reason, there is no conclusory evidence in the prior art which explains the exact corrosion mechanism which produces the nodular corrosion. This limits, to some degree, the capability to ascertain whether other alloys will be susceptible to nodular corrosion before actually placing samples made from these alloys into reactors.
Laboratory tests conducted under the conditions normally experienced in a reactor at approximately 300.degree. C. and 1000 psig in water, but absent radiation, will not produce a nodular corrosion product on Zircaloy alloys like that found in some instances on Zircaloy alloys which have been used in reactor service. However, if steam is used, with the temperature increased to over 500.degree. C. and the pressure raised to 1500 psig, a nodular corrosion product like that occasionally found on Zircaloy in reactor service can be produced on Zircaloy alloys in laboratory tests. Specimens of Zircaloy alloys which are annealed at 750.degree. C. for 48 hours are particularly susceptible to nodular corrosion under these test conditions. These annealed Zircaloy specimens will produce, in tests run for relatively short times, i.e. 24 hours, a degree of nodular corrosion comparable to that of Zircaloy tube cladding in actual reactor service that has been found to have nodular corrosion. At this higher temperature and pressure, a simulated nuclear reactor environment is provided which will allow researchers to determine the susceptibility of new alloys to nodular corrosion. With this test, a comparison between samples from new alloys and Zircaloy specimens tested under the same conditions can be made.
To be considered as a suitable alternate or replacement for the Zircaloy alloys, any new alloy must not only be less susceptible than the Zircaloy alloys to nodular corrosion, but must maintain acceptable uniform corrosion rates, comparable to those of the Zircaloy alloys, to ensure sufficient service life. Zircaloy alloys have been used extensively as fuel rod cladding and are known to contain many desirable properties that alternate or replacement alloys must also contain. Zircaloy alloys have the desirable properties of a low neutron absorption cross section and at temperatures below 750.degree. F. are strong, ductile, extremely stable and as mentioned previously have excellent uniform corrosion resistance in water at reactor operating temperatures.
Fuel element performance has revealed another problem with brittle splitting of nuclear fuel element cladding due to the combined interactions between the nuclear fuel, the cladding and the fission products produced during nuclear fission reactions. It has been discovered that this undesirable performance is due to localized mechanical stresses on the fuel cladding resulting from differential expansion and friction between the fuel and the cladding. Fission products are created in the nuclear fuel by the fission chain reaction during operation of a nuclear reactor, and these fission products are released from the nuclear fuel and are present at the cladding surface. These localized stresses and strains in the presence of specific fission products, such as iodine and cadmium, are capable of producing cladding failures by phenomena known as stress corrosion cracking or liquid metal embrittlement.